Evaluation of activated area in the electrostatic accelerator facilities

In order to clear the activated area in electrostatic accelerator facilities,four accelerator facilities were selected and typical neutron emission experiments were performed.Neutron flux during operation and induced activity caused by charged particles on the accelerator and its surrounding area after irradiation were measured.Also the monitored neutron flux and calculated value by Monte Carlo calculation using PHITS code were compared.It was confirmed that the results between calculated data and measured data showed the good agreement with each other.Finally,it was concluded that we have to take care the activation of beam line and target.But,it is not necessary to treat accelerator tank,surrounding materials,and building concrete as radioactive materials in case of decommissioning.     

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

Atomic displacement cross sections for carbon from ENDF/B-III data

Atomic displacement cross sections have been derived from the data in materials file 1165 (Carbon-12) of the third version of the Evaluated Nuclear Data File (ENDF/B-III), Using the Lindhard theoretical treatment for the stopping cross sections of carbon ions in carbon. A comparison of the differences between previous estimates of the carbon atom displacement functions and the results of these calculations indicate that the use of the atomic displacement cross sections presented herein should improve the correlations of damage production in different irradiation facilities.     

结构材料核的原子位移截面计算

中子辐照引起的材料损伤是裂变反应堆设计的重要考虑因素。对于晶体结构材料,其辐照损伤主要来自晶格原子的位移。结构材料核与中子发生带电粒子反应的截面、原子位移(DPA)截面、KERMA因子是计算辐照损伤的基础。为比较不同程序计算的DPA截面的差异和基于不同评价核数据库的DPA截面的差异,采用核模型计算程序UNF及核数据处理程序NJOY计算了27 Al、48 Ti、90Zr、Cr、Fe、Ni、Cu等结构材料核的DPA截面,将二者计算结果进行了比较分析;比较分析了基于不同评价核数据库的采用NJOY计算的DPA截面;比较分析了NJOY与蒙特卡罗程序计算的DPA截面。结果表明,UNF与NJOY的结果存在一定的差别,不同评价库的结果也是有差别的,蒙特卡罗程序采用不同模型计算时结果也存在一定的差别。     

Neutron damage calculations in Cu,Nb and Au to 32 MeV: Application to sputtering and deuteron-breakup neutron sources

Primary recoil distributions and specific damage energies have been computed for high energy deuteron-breakup neutrons in Cu, Nb and Au. The calculations are based on theoretical neutron cross sections and consider in particular a d-Be spectrum broadly peaked at 15 MeV with some neutrons above 30 MeV. The theoretical results are similar to corresponding calculations for monoenergetic 15-MeV neutrons and are in good agreement with range measurements of (n, 2n) recoils generated by high energy d-Be neutrons in Nb and Au. The calculations are also consistent with recent d-Be neutron sputtering experiments in Nb and Au and demonstrate the usefulness of deuteron-breakup neutron sources for simulating fusion neutron effects.     

核数据处理软件NECP-Atlas中的光子相关数据计算方法研究

NECP-Atlas是西安交通大学自主研制的核数据处理软件,具有丰富的功能,可将评价核数据制作为后续核设计所需的应用核数据库,本文在NECP Atlas中建立了光子相关数据的计算方法,可计算产生中子核反应释放的瞬发光子产生截面、光子与原子的反应截面,裂变产物衰变释放的缓发光子多群产生矩阵,以及光子辐照损伤截面等数据。数值结果显示,如果不考虑缓发光子,钠冷快堆中控制组件、反射层组件的光子功率与参考解的最大偏差可达3258%、2041%,采用NECP Atlas计算的多群缓发光子产生矩阵后两类组件偏差降为093%以下。采用文献结果对Fe的光子辐照损伤截面进行了验证,计算结果与参考解吻合良好。     

Displacement per atom calculation in YBCO superconductors through Monte Carlo simulation

The results of the calculations of the displacements per atom distribution induced by the gamma irradiation up to 15 MeV on YBa₂Cu₃O_7−x superconducting slabs are presented. Firstly, a calculation procedure for the displacement cross sections and the displacement per atom distributions was applied using the Monte Carlo simulation through the MCNPX code system. Then, based on this algorithm, the displacement per atom in-depth distributions were calculated starting from the energy flux distributions obtained from the simulation process, taking into account the contribution from each atom, obtaining a predominance of the Cu-O₂ planar sites over yttrium and barium atoms and more specifically the oxygen atoms predominate at low energies and the copper atoms at higher energies. Finally, the linear correlation observed between the displacement per atom distributions and energy deposition profiles at each incident energy was analyzed.     

Neutron Self-Shielding Factor of Scattering-Predominant Resonance Foils

The neutron self-shielding factor of ⁵⁹Co resonance foil as an example of foils whose scattering cross section predominate over their absorption cross sections was obtained by both Monte Carlo method (analog) and the collision probability method for various thicknesses of the foil. Also, the transmission and reflection probabilities of neutrons which have various energies near the resonance energy were obtained, and the effects of multiple scattering of neutrons on the neutron self-shielding factor are discussed.

The neutron self-shielding factors obtained by the Monte Carlo method and by the collision probability method agreed well with each other in the cases Σ t ～ 4.0, in which the Monte Carlo method requires considerably longer machine time. Although for the cases of large Σ t (～4.0) the agreement is not always good because of the flat flux approximation in the collision probability method, the calculation time by Monte Carlo is conveniently short. A combination of both methods is useful in obtaining the neutron self-shielding factor of resonance foils.     

The primary recoil spectrum in the simulation of fast-reactor radiation damage by charged-particle bombardment

The energy spectrum of primary knock-on atoms in iron subject to a typical fast-reactor neutron spectrum is calculated and compared to the PKA.spectrum calculated in nickel under bombardment by 46.5 MeV nickel ions (as on the Harwell VEC), 5 MeV nickel ions, 22 MeV carbon ions, 5 MeV protons and 1 MeV electrons. The damage produced by each PKA spectrum is calculated, and the proportion of the damage produced by high-energy recoils estimated. Comparison of the PKA spectra indicates that the nickel-ion bombardments are the closest simulation of fast-reactor damage.     

蒙特卡罗方法在中子能谱研究中的应用

针对快中子能谱实验测量数据的修正 ,开发了FAMS MC蒙特卡罗计算程序 ,对 5 9、6 4和1 4 1MeV中子在Be核上产生的次级中子双微分截面的实验结果进行了修正计算 ,并与用MCNP蒙特卡罗程序修正的结果进行了比较。用FAMS MC程序进行中子能谱实验测量数据修正得到满意结果。     

Determination of the neutron flux and energy spectrum in the low-temperature fast-neutron facility in CP-5, calculations of primary-recoil and damage-energy distributions,and comparisons with experiment

We have determined the absolute differential neutron-energy spectrum for the low-temperature fast-neutron irradiation facility in the CP-5 reactor by means of a 20-foil activation technique. This technique employs the most recent version of the SAND-II computer code, which iteratively unfolds the neutron spectrum by fitting the foil activities. A Monte Carlo routine was also employed to calculate standard-deviation errors in each neutron-energy group. Using this differential neutron spectrum we have calculated, for numerous elements, total recoil cross sections, detailed primary-recoil group distributions, total damage-energy cross sections, damage-energy distributions, and an error analysis based on the uncertainties in the neutron spectrum. The significance of this information with respect to the interpretation of various neutron radiation-damage experiments, including sputtering, disordering of ordered alloys, and changes in critical current of A-15 compound superconductors is discussed. A detailed comparison is made among initial resistivity-damage rates for five widely different (well characterized) neutron sources, fission fragments, and heavy ions.     

Monte Carlo Simulations of 17.9–22.3 MeV Energetic Proton Irradiation Effects on bcc-Zirconium Fusionic Materials

Advanced fusion structural materials (FSMs) have impact role in terms of efficiency of nuclear energy production. Besides engineering and design of fusion reactors, radiation durability of FSMs is another valuable issue that cannot be ignored. 17.9–22.3 MeV proton irradiation of bcc-Zirconium Fusion Structural Material was evaluated by using Monte Carlo based simulation tools. Total binary reaction cross sections were respectively calculated as 1167.6 and 1273.92 mb for 17.3- and 22.3 MeV proton energies via TALYS-1.6 version. Additionally, residual production cross sections and total particle production cross sections were obtained and analyzed by the TALYS code. Radiation damage parameters as Displacement Per Atom (DPA) and Stopping Power (SP) were studied by SRIM-2013 version. FLUKA 2011.1 used for only DPA calculations and making a complete comparison with the other calculation results. SP and Number of Secondaries were found by using GEANT4.10.p.04 version simulations. Natural Zr(p,x) reactions were studied in the given energy values in the plane of reaction probability and radiation damage calculations.     

Stopping cross sections of liquid water for MeV energy protons

Cross sections for the stopping of swift protons in liquid water have been measured for the first time by using a liquid water jet target of 50 μm in diameter. The energy loss spectra of incident 2.0 MeV protons were measured at scattering angles of 5.0-50 mrad. Experimental energy loss spectra have been successfully reproduced by Monte Carlo simulation calculations (GEANT4.9.1.p02 toolkit) by taking account of multiple scattering of projectile ions inside the liquid water target. The present stopping cross sections are found to be considerably smaller than other standard stopping power data, revealing e.g. about 11% deviation from those of SRIM2003.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Irradiation damage to the beam window in the 800MWth accelerator-driven system

Irradiation damage to the beam window in the concept of 800MWth accelerator-driven system is evaluated. Heat produced in the window is also evaluated. Transport of proton and neutron up to 3.0 GeV is calculated by both PHITS that is the Monte Carlo code for particles and heavy ions and TWODANT that is two-dimensional deterministic transport code. The beam window is irradiated at the center of the accelerator-driven system with 20 MW proton beam power and neutron from the core during 300 full power days. Heat, displacement per atom, production rate of hydrogen and helium isotopes, and neutron and proton fields are estimated, assuming the Gaussian and flat beam profiles.     

Calculated neutron cross sections for Cu and Nb up to 32 MeV for neutron damage analysis

Cross sections for neutron interaction with Cu and Nb, with emphasis on spectra of light particles from binary reactions, are calculated for neutron energies from 4 to 32 MeV for estimating recoil probability densities for the analysis of damage experiments with a Be (d, n) neutron source. Nuclear model parameters were adjusted to reproduce the available cross-section data around 14 MeV. Helium production cross sections were also calculated for ⁶³Cu for neutrons below 20 MeV, as an illustration of the Hauser-Feshbach method for calculating tertiary reaction cross sections.     

基于OpenMC的多群截面库制作及有效性验证

OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。     

Feasibility study of photo-neutron flux in various irradiation channels of Ghana MNSR using a Monte Carlo code

Computer simulation was carried out for photo-neutron source variation in outer irradiation channel, inner irradiation channels and the fission channel of a tank-in-pool reactor, a Miniature Neutron Source Reactor (MNSR) in sub-critical condition. Evaluation of the photo-neutron was done after the reactor has been in sub-critical condition for three month period using Monte Carlo Neutron Particle (MCNP) code. Neutron flux monitoring from the Micro Computer Control Loop System (MCCLS) was also investigated at sub-critical condition. The recorded neutron fluxes from the MCCLS after investigations were used to calculate the power of the reactor at sub-critical state. The computed power at sub-critical state was used to normalize the un-normalized results from the MCNP.     

Characteristics of Fabricated Neutron Detectors Based on a SiC Semiconductor

Wide-band-gap semiconductors such as SiC, AlN, and GaN are promising materials for harsh environment applications due to their high-temperature operation capability. Two types of PIN-type semiconductor neutron detectors based on SiC were designed and fabricated for nuclear power plant (NPP) applications such an in-core reactor neutron flux monitoring and safeguarding nuclear materials. One is for fast neutron detection and the other, which was evaporated with ⁶LiF, is for thermal neutron detection. In this study, preliminary tests, such as the determination of I-V and alpha responses, were performed. Reaction probabilities with respect to neutron energies were also calculated by using an MCNPX code for comparison with the experimental results. Responses of the neutrons were measured at the Ex-core Neutron irradiation Facility (ENF) of the High-flux Advanced Neutron Application Reactor (HANARO) research reactor at the Korea Atomic Energy Research Institute (KAERI). Pulse height spectra and count rates were measured with respect to the neutron fluxes from 1:6 × 10⁶ n/cm²·s to 1:9 × 10⁷ n/cm²·s. Also, a 0.99 root-mean-square value of linearity against the fluxes to the count rates was obtained with the fabricated neutron detectors. For a thermal neutron detector, a 3.3% detection efficiency was obtained.     

Investigation of the projectile atomic number influence to the total sputtering yield in the keV energy region

The non-monotonous dependence of the total sputtering yield on the projectile atomic number, which is unexpected in the frame of the Sigmund linear cascade theory, is investigated using Monte Carlo simulations (program SRIM 2003). This effect is studied on the example of aluminum sputtering by six different projectiles (N, Ne, Al, Ar, Kr and Xe) at normal incidence. The incident projectile energy is 2 keV. Investigation consists of the analyses of ASI distributions of sputtered atoms as well as of nuclear energy loss depth distributions of projectiles with fixed number of ejected atoms. The results show that the non-monotonous behavior of Y(Z) is due to the ability of projectiles somewhat lighter than aluminum to efficiently eject large number of atoms by formation of collision cascades in the subsurface region which are directed towards the surface. On the other hand, ions that are heavier or significantly lighter than aluminum cannot form this type of cascades - the heavier ions cannot transfer a lot of energy to recoils in a primary knock-on collision that will move towards the surface, while significantly lighter ions transfer the energy too deep into the target.